1. Field of the Invention
This invention relates to pressurized water reactors and more particularly to responses to the rupture of steam generator tubes using passive safety systems without the need for intervention by an operator.
2. Background Information
In a pressurized water reactor (PWR) nuclear power plant, reactor coolant in the form of light water is circulated through the core of a reactor where it is heated by controlled fission reactions in fuel contained within fuel assemblies making up the core. The heated coolant is circulated within containment in a primary circuit through a hot leg to a steam generator where it passes through heat exchanger tubes and then returns to the core through a cold leg. Feed water covering the heat exchanger tubes in the steam generator is converted into steam which is circulated in a secondary circuit to a turbine generator outside of containment which uses the steam to generate electricity.
If a leak develops in one of the thousands of heat exchanger tubes in the steam generator, primary coolant leaks into the feed water in the secondary side of the steam generator. This is a compromise of the primary-secondary loop barrier and has several adverse consequences. First, it results in a loss of coolant in the reactor coolant system. The PWR includes a pressurizer which is a tank containing reactor coolant and a head of steam which maintains pressure in the primary loop. For small steam generator tube ruptures, the pressurizer can maintain the pressure in the primary loop for a period of time.
Conventionally, the PWR also has a chemical and volume control system which regulates the concentration of moderators in the reactor coolant and provides makeup coolant. In the event of a moderate steam generator tube rupture, the chemical and volume control system can make up for the loss of reactor coolant. However, this system adds coolant to the primary loop under pressure so that coolant is continually lost through the rupture.
PWRs also have safety injection systems which inject coolant into the reactor should the pressure drop below a predetermined level. Pressurized water reactors further have a refueling water supply tank used for refueling but also available to flood the reactor core in the event of a severe loss of coolant accident (LOCA).
Another adverse affect of a ruptured steam generator tube is that it floods the secondary side of the steam generator. Automatic controls which regulate the level of feed water in the steam generator terminate the flow of feed water in the event primary fluid leaks into the secondary side of the steam generator. However, even with the flow of feed water terminated, if the leak is sufficient, primary coolant can completely flood the secondary side of the steam generator and overflow into the steam line. This can result in a radiation leak through pressure relief valves provided in the secondary circuit.
Currently, in the event of a rupture of a steam generator tube, the plant operators are required to perform a number of actions. First, they must identify that the event has occurred. This can be accomplished by observing relative steam generator feed water levels between two or more steam generators in the plant. The steam generator in which the feed water rises faster than the others is the faulted unit. Also radiation levels in the steam generator blow-down lines can help identify the condition. The operator must isolate the faulted steam generator by closing the main steam line isolation valve for that steam generator. The reactor is then cooled down using the intact steam generator by reducing steam side pressure. The operator either reduces the pressure set point of the steam bypass controller which dumps steam to a condenser, or the operator reduces the set point of the steam generator power operated relief valve(s). Once the reactor is cooled down sufficiently (about 50.degree. F.), it is possible to reduce the reactor pressure to equal that of the isolated faulted steam generator which will terminate the leak. Normally, this will be done by the operator opening a pressurizer spray valve which will condense some steam in the pressurizer thereby reducing the pressure.
The normal response of the previously mentioned safety injection system is to provide higher pressure injection during the postulated event. However, once the reactor pressure is reduced to equal the faulted steam generator pressure (typically about 1,100 psig), the high pressure safety injection pumps will naturally increase their injection and tend to repressurize the reactor. To prevent this repressurization, which would restart the steam generator tube leak, the safety injection pumps have to be stopped. The operators are required to carefully check the reactor conditions to ensure that it is safe to stop the high pressure injection pumps.
The above procedures are designed for responding to a rupture in a single steam generator tube. Currently, regulators are also interested in plant performance in the event of multiple steam generator tube ruptures, namely, in the range of three to seven tube ruptures. In current plants, such an event might result in over filling of the faulted steam generator because of the very rapid operator response that would be required.
Currently, there are passive safety systems under development for PWRs. Such passive systems do not rely upon active components such as pumps and do not require operator action. U.S. Pat. No. 4,753,771 assigned to the assignee of the present invention is directed to passive safety systems for PWRs. One such system is a passive heat removal system which utilizes a heat exchanger immersed in the cold coolant in the refueling water supply tank and connected between the hot and cold legs of the primary circuit. A normally closed valve prevents flow through this heat exchanger under normal conditions. When the temperature of the reactor coolant reaches a predetermined value, such as would occur for example if the steam generator was not providing sufficient heat removal, the valve is opened and reactor coolant circulates by convection through the heat exchanger to remove decay heat from the coolant. The PRHR HX may alternatively be actuated when the SG water level drops to a predetermined level.
Another passive safety system described in U.S. Pat. No. 4,753,771 is a core make-up tank. The core make-up tank(s) contain cold reactor coolant and are pressurized by steam provided through a steam line from the pressurizer. A normally closed valve isolates the core make-up tank(s) from the primary loop. If the liquid level in the pressurizer drops below a predetermined level, the valve is opened to connect the core make-up tank(s) to the primary system. The core make-up tank is mounted relative to the pressurizer so that under these conditions, the liquid level in the core make-up tank will be sufficient such that the core make-up tank will add coolant to the primary loop. This would make up for minor losses of reactor coolant in the event of a steam generator tube rupture; however, this system does not respond until the level in the pressurizer has dropped to the predetermined level.
There is a need for an improved apparatus and method for responding to a steam generator tube rupture to quickly terminate the leak while assuring proper continued cooling of the reactor core, and particularly an apparatus and method which utilizes passive safety systems and does not require the intervention of an operator.